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Ted1nicallnfoonation Bulletin - Maximum Internal Dose

Estimates for Certain DOE Complex Claims

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Effective Date:O1/10/2004

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Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 3 of 20 

MAXIMUM INTERNAL DOSE ESTIMATES FOR CERTAIN DOE COMPLEX CLAIMS 

1.0 PURPOSE 

The purpose of this Technical Information Bulletin (TIB) is to provide a method to facilitate timely 

processing of claims under the Energy Employee Occupational Illness Compensation Program Act 

(EEOICPA 2000), which involve cancer to an organ with little or no reported internal dose from 

internally deposited radionuclides that might be associated with work at DOE complex sites. 

2.0 BACKGROUND 

Title 42, Part 82 of the Code of Federal Regulations (CFR) dictates the methods to be used for 

radiation dose reconstruction under the Energy Employees Occupational Illness Compensation 

Program Act of 2000. Section 82.10(k) summarizes the general philosophy to be adopted: 

“Research and analysis will be determined sufficient if one of the following three 

conditions is met: 

(1) From acquired experience, it is evident the estimated cumulative dose is sufficient 

to qualify the claimant for compensation (i.e., the dose produces a probability of 

causation of 50% or greater); 

(2) Dose is determined using worst case assumptions related to radiation exposure 

and intake, to substitute for further research and analyses; 

(3) Research and analysis indicated under steps described in paragraphs (f) - (j) of this 

section have been completed. 

Worst-case assumptions will be employed under condition 2 to limit further research 

and analysis only for claims for which it is evident that further research and analysis will 

not produce a compensable level of radiation dose (a dose producing a probability of 

causation of 50% or greater), because using worst-case assumptions it can be 

determined that the employee could not have incurred a compensable level of radiation 

dose.” 

“Worst-case assumption” is defined in Section 82.5(r) as: 

“A term used to describe a type of assumption used in certain instances for certain 

dose reconstruction conducted under this rule [42 CFR 82]. It assigns the highest 

reasonably possible value, based on reliable science, documented experience, and 

relevant data to a radiation dose of a covered employee.” 

A number of EEOICPA claims have been submitted that can be expedited based on paragraph (2) 

above. Cases were reviewed to identify those with cancer of an organ that does not concentrate any 

of the primary radionuclides expected to be found at DOE sites and those with little or no apparent 

internal dose to that organ. 

In accordance with OCAS-IG-002 (Internal Dose Reconstruction Implementation Guideline, Rev. 0, 

August 2002), internal dose is assigned to employees who were monitored but had no detectable 

activity (“positive”) in their samples and to employees who were not included in a bioassay program, 

because there is some amount of intake and associated dose that is not detectable by an internal 
Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 4 of 20 

dosimetry program. For employees who were monitored and who do have detectable activity 

(“positive”) in their samples, those sample results will be evaluated to ensure the method described in 

this TIB results in a greater probability of causation. To expedite the dose reconstructions, cases that 

met the criteria above will first be evaluated with the method described in this TIB. If the outcome 

yields a probability of causation >50%, a dose reconstruction using more reasonable assumptions will 

be performed. 

3.0 IMPLAUSIBLE UNDISCOVERED INTAKES 

3.1 ASSUMPTIONS 

For claims where it is considered likely that the covered employee had no significant internal radiation 

exposure, a method to expedite claims has been developed in accordance with 42 CFR 82.10(k)(2). 

This method assumes the “largest reasonably possible value” of the source term comprised of 

radionuclides that are/were typically the more significant radionuclides (by either preponderance or by 

internal dose significance) on a site. For this “worst-case” estimate of internal dose, it is assumed that 

on the first day of the first year of employment, the covered employee had an acute inhalation intake 

of each of the radionuclides in the source term, in the amounts listed below. 

Based on historical data, it is believed to be highly unlikely that such an intake could have occurred 

without being detected by workplace monitoring at the time. It is also believed that this is a significant 

overestimate of internal dose for an unmonitored covered employee or a covered employee with no 

internal monitoring above detection thresholds. 

3.1.1 Additional Assumptions 

Additional assumptions to develop this method are: 

•	 All intakes are inhalations of standard 5 micrometer AMAD, except for I-131, which is assumed 

to be in vapor form (class SR-1). 

•	 The most soluble form of the radionuclide specified in ICRP 1994a was used to maximize 

dose to systemic organs, except as noted below; dose to lung is not germane to this exercise. 

•	 Because maximum permissible body burdens were the metric (actually uptake) for so many 

years, the assumed implausible uptake was based on a percent of the radionuclide-specific 

Maximum Permissible Body Burden (MPBB) for soluble chemical forms as defined by the 

National Committee on Radiation Protection (NCRP1959). It was assumed that an intake 

resulting in 10% of a MPBB would not have likely occurred to an unmonitored worker or would 

have likely resulted in a readily noticeable bioassay result in a monitored worker, readily 

noticeable air sample, or other indicators of personnel contamination. In other words, an event 

providing the possibility of an intake resulting in a body burden exceeding 10% of the MPBB 

would not have gone unnoticed and there would be some sort of indication in the worker’s 

records. This assumption applies to bona fide DOE sites and national labs with active 

radiation protection programs, not to Atomic Weapons Employers (AWEs). The current ICRP 

methodology is used calculate doses from these implausible intakes. 

•	 For type F and M materials the associated derived intake (i.e., intake resulting in a 10% 

MPBB) was assumed to be 10 and 20 times the 10% MPBB, respectively. The factors of 10 

and 20 come from the current ICRP models and the differences between an intake and the 

activity that is present in the body after the initial clearance of the short-term compartments. 
 

 

Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 5 of 20 

These factors are used to relate the historical quantity of control, body burden, which was 

based on ICRP 2 methods to the present quantity of control, intake, which is based on current 

ICRP methods. These factors were estimated from tables in the November 2002 issue of 

Health Physics that give the intake retention fraction for the whole body (without the ET region) 

as a function of time after acute intake for different elements and inhalation types. Because 

initial deposition in the lung was usually not considered by the ICRP (1959) to be part of the 

“body burden,” the retention fractions used allowed some time for the rapid clearance 

components. Examples are listed in Table 3.1.1-1. 

Table 3.1.1-1. Fractional retention in whole body after initial clearance from lung.a 

Element 

Fractional retention in whole body after initial clearance from lung 

Inhalation type F Inhalation type M 

3 days post intake 4 days post intake 60 days post intake 90 days post intake 

Strontium 0.227 0.119 

Cesium 0.449 0.440 

Cerium 0.0683 0.0654 

Ruthenium 0.251 0.227 0.0392 0.0392 

Barium 0.147 0.104 

Lanthanum 0.307 0.291 0.0683 0.0654 

Zirconium 0.278 0.252 0.0496 0.0448 

Niobium 0.0476 0.0409 

Cobalt 0.0410 0.0337 

Technetium 0.217 0.160 0.0321 0.0238 

Europium 0.0605 0.0568 

a. Source: Potter 2002 

•	 The assumption of type S for Co-58 and Co-60 is used because it results in larger doses to the 

systemic organs because of the high-energy photons. Although the logic in the bullet above 

does not directly apply to type S material, the fractional retention in the whole body is similar 

for type M and type S at 60 and 90 days, so the derived intake is estimated as 20 times the 

10% MPBB. 

•	 Zn-65 is classified as type S in ICRP 68. As for Co-58 and Co-60 type S materials, the logic of 

determining a body burden to intake conversion factor does not directly apply, but use of the 

factor of 20 does not seem unreasonable. 

•	 Mn-54 type M has a larger dose conversion factor for most organs/tissues and was generally 

more claimant favorable than type F. 

•	 Intakes of tritium are not considered by this method and must be addressed by site, depending 

on the site-specific detection level. Tritium clears very rapidly from the body so multiple 

undetected intakes are possible, although H-3 is readily detected at very small intake levels by 

urinalysis. Tritium dose was often reported with an individual’s external dose prior to 1989; in 

such cases no further H-3 analysis may be necessary. 

•	 To be generic to most of the DOE sites, intakes were assumed to involve the most plausible 

radionuclides for all the sites, even though it is implausible that any one worker had intakes of 

all the radionuclides. Two groups of radionuclides were considered: sites having one or more 

reactors and sites without a reactor. The Nevada Test Site was considered a special case for 

which these two groups of radionuclides might not be appropriate. A review indicated that 

after 1971, the source term at NTS, except in the tunnels, was consistent with that chosen for 

reactor sites. 
 

Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 6 of 20 

•	 For sites without a reactor, the actual ratio of radionuclides in mixtures is not relevant because 

each radionuclide is used at its maximum amount. For sites with reactors, the fission products 

were chosen for the list based on their relative abundance in fuel or contamination after 180 

days of post-irradiation cooling time. This is very conservative for any contamination at 

facilities other than the reactors themselves. 

•	 A review of the assumed intakes from the method described above and associated possible 

air concentrations for different periods indicate that the uranium intakes based on soluble 

uranium may not have been sufficiently conservative for uranium facilities that controlled their 

programs based on consideration of uranium exposure alone. The derived intakes for 

uranium-234 and uranium-238 were increased by a factor of 100 to ensure that this scenario 

will bound doses to those employees who worked at uranium facilities, but were not included 

in a bioassay program or had no detectable occupationally related activity in them or their 

samples. Derived intake values for U-234 and U-238 respectively may be lowered to 50 nCi 

and 5 nCi for employees who worked in areas that were controlled based on the possible 

presence of other alpha emitters, such as plutonium-238 or plutonium-239. 

After the intakes of the radionuclides were determined, the alpha radionuclides and the beta/gamma 

radionuclides were each summed to show the total intake of each for the reactor and non-reactor 

sites. These alpha and beta/gamma intakes were then used to consider what the average air 

concentrations would have been if the intakes had occurred chronically rather than acutely. A 

breathing rate of 2.4 E+09 milliliters per year, based on Reference Man, was used to calculate the air 

concentrations for 1, 10 and 30 years that would produce the given intakes. Site air concentration 

criteria (area controls, respirator usage, etc.) were based on the permissible air concentration of 

radionuclides assumed to be present in significant quantities. At many facilities, air concentration 

controls were specified for alpha, based on Pu-238/239/240, and beta/gamma, based on Sr-90. The 

typical air concentration controls listed in Table 3.1.1-3 are based on the radionuclide air 

concentration values cited in NCRP (1959), AEC (1968), and 10 CFR 835, which for most 

radionuclides did not change significantly over time. In later years, the beta/gamma “limit” may have 

increased to 2 E-9 uCi/ml, but the control criteria also changed and were based on 10% of the limit, 

which would equate to an air concentration of 2 E-10 uCi/ml. 

It can be seen from the Table 3.1.1-3 that even chronic intakes at these levels would exceed typical 

air concentration controls that were being used at most sites since the 1950s to manage access to 

and work practices in areas that had increased radioactivity in air. Acute intakes are assumed for this 

worst-case approach, because these will result in larger assignments of dose. 

Table 3.1.1-3 also shows that workers who spent significant time in areas where uranium air 

concentration controls were used to manage access and work practices do not appear to be good 

candidates for the non-uranium facility approaches. The non-uranium facility approaches can be used 

for reconstruction of doses for claims that did not involve exposure in uranium process environments 

for more than a few years. The uranium intakes are included with the alpha intakes, because it is 

unlikely that the measured alpha air concentration would have excluded uranium. Uranium is shown 

by itself for nonreactor facilities, because the air concentration control in a uranium facility was likely 

to be larger than the cited gross alpha air concentration control. For facilities with reactor source 

terms the uranium air concentration was not listed, because it is believed that the control would have 

been on the gross alpha air concentration and because the uranium concentration will be the same as 

the comparable nonreactor facility. 
 

 

Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 7 of 20 

Table 3.1.1-2. Radionuclides of interest and respective maximum permissible body burdens, 

assumed intake absorption type, derived maximum plausible intake quantities, and absorption type, 

radiation type and IMBA type and version used for estimating organ doses at DOE sites. 

Radionuclide 

Soluble (S) 

MPBB 

(nCi) a Assumed intake 

absorption type 

Derived intakeb 

used in IMBAc 

runs (nCi) 

Absorption 

type used in 

IMBAc

 runs Radiation type 

IMBAc 

type 

IMBAc 

version 

Sites without a reactor: 

Sr-90 2,000 F 2,000 F Electron > 15 keV OCAS 3.0.48 

Tc-99 10,000 F 10,000 F Electron > 15 keV OCAS 3.0.63 

Cs-137 30,000 F 30,000 F Electron > 15 keV NIOSH 1.0.42 

U-234 50 F 5,000d F Alpha NIOSH 1.0.42 

U-238 5 F 500d F Alpha NIOSH 1.0.42 

Am-241e

 50 M 100 M Alpha NIOSH 1.0.42 

Cm-244 100 M 200 M Alpha OCAS 3.0.48 

Np-237 60 M 120 M Alpha OCAS 3.0.48 

Pu-238 40 M 80 M Alpha NIOSH 1.0.42 

Pu-239 40 M 80 M Alpha NIOSH 1.0.42 

Th-230f

 50 M 100 M Alpha OCAS 3.0.63 

Cf-252 40 M 80 M Alpha OCAS 3.0.63 

Sites with a reactor–all those above plus: 

Mn-54 20,000 M 40,000 M Photon>250 keV OCAS 3.0.63 

Co-58 30,000 M 60,000 S Photon>250 keV OCAS 3.0.63 

Co-60 10,000 M 20,000 S Photon>250 keV NIOSH 1.0.42 

Fe-59 20,000 F 20,000 F Electron > 15 keV OCAS 3.0.63 

Zn-65 60,000 M 120,000 S Photon>250 keV OCAS 3.0.63 

Y-91 5,000 M 10,000 M Electron > 15 keV OCAS 3.0.63 

Nb-95 40,000 M 80,000 M Electron > 15 keV OCAS 3.0.59 

Zr-95 20,000 F 20,000 F Electron > 15 keV OCAS 3.0.59 

Ru-103 20,000 F 20,000 F Electron > 15 keV OCAS 3.0.59 

Ru-106 3,000 F 3,000 F Electron > 15 keV OCAS 3.0.59 

I-131 700 F 700 F Electron > 15 keV OCAS 3.0.63 

Ce-141 30,000 M 60,000 M Electron > 15 keV OCAS 3.0.63 

Ce-144 5,000 M 10,000 M Electron > 15 keV OCAS 3.0.63 

Pm-147 60,000 M 120,000 M Electron > 15 keV OCAS 3.0.63 

Eu-154 5,000 M 10,000 M Electron > 15 keV OCAS 3.0.59 

Eu-155 70,000 M 140,000 M Electron > 15 keV OCAS 3.0.59 

a. Source: NCRP (1959). 

b. Derived maximum plausible intake based on 10% of the soluble NCRP (1959) MPBB multiplied by factors of 10 and 20 for assumed 

intake absorption types of F and M, respectively. 

c. Source: Birchall 2002, 2003, 2003a, and 2003b. 

d. The uranium values for the general approach are increased to assure that this is a bounding approach for all facility-types. The lower 

values of 50 nCi for U-234 and 5 nCi for U-238 can be used to assign “worst-case” doses to individuals who did not work in areas 

where controls were based primarily on uranium exposure concerns. 

e. Pu-241 is accounted for by the Am-241 intake value. 

f. Th-230 can be used as a surrogate for Th-232. 

Table 3.1.1-3. Total intakes and possible average air concentrations. 

Facility type Radiation type 

Intake 

(uCi) 

Air concentration 

based on 1 year 

intake (uCi/ml) 

Air concentration 

based on 10 year 

intake (uCi/ml) 

Air concentration 

based on 30 year 

intake (uCi/ml) 

Typical air 

concentration 

controls (uCi/ml) 

Non-reactor 

and nonuranium 

Alpha 0.815 3.4E-10 3.4E-11 1.1E-11 2 E-12 

Beta/gamma 42 1.8E-08 1.8E-09 5.8E-10 3 E-10 

Uranium 0.055 2.3E-11 2.3E-12 7.6E-13 7 E-11 

Reactor and 

non-uranium 

Alpha 0.815 3.4E-10 3.4E-11 1.1E-11 2 E-12 

Beta/gamma 776 3.2E-07 3.2E-08 1.1E-08 3 E-10 

Non-reactor 

and uranium 

Alpha 6.26 2.6E-09 2.6E-10 8.7E-11 2 E-12 

Beta/gamma 42 1.8E-08 1.8E-09 5.8E-10 3 E-10 

Uranium 5.5 2.3E-09 2.3E-10 7.6E-11 7 E-11 

Reactor and 

uranium 

Alpha 6.26 2.6E-09 2.6E-10 8.7E-11 2 E-12 

Beta/gamma 776 3.2E-07 3.2E-08 1.1E-08 3 E-10 
Effective Date: 01/10/2004 Revision No. 01 PC Procedure No. ORAUT-OTIB-0002 Page 8 of 20 

To encompass employees who may have worked in facilities where uranium was the primary 

radionuclide of concern for internal dose, the assumed uranium intakes were each increased by a 

factor of 100. The factor of 100 was chosen to assure that this approach adequately bounds the level 

of exposure that potentially could have gone unnoticed at uranium facilities. 

To facilitate use of these data, two spreadsheets were created based on annual organ doses 

calculated with IMBA (Birchall 2002, 2003, 2003a, and 2003b) for the derived acute inhalation intakes 

shown in Table 3.1.1-2 and for the same derived intakes, except that the U-234 and U-238 intakes are 

lowered respectively to 50 and 5 nanocuries for non-uranium facilities. These spreadsheets allow the 

user to input claim specific data to obtain data necessary for the IREP program for the organs shown 

in Table 3.1.1-4. This approach may be also used for diagnosed conditions that are to have internal 

doses calculated using largest dose to a non-modeled organ as specified in OCAS-TIB-008. 

Table 3.1.1-4. Organs appropriate for applying the maximum internal dose for 

certain DOE complex claims. 

Adrenals Esophagus Pancreas Thymus 

Brain G. Bladder Skin U. bladder 

Breast Heart Spleen Uterus 

Colon Muscle Stomach 

4.0 CALCULATION OF INTERNAL DOSE FROM MAXIMUM INTAKES 

The calculation of organ dose assumes that the covered employee had maximum intakes of all of the 

radionuclides listed in Table 3.1.1-1. While individual exposures to a single radionuclide from the 

mixtures defined in this method possibly could be underestimated, despite the attempt to overestimate 

each radionuclide intake, it is believed that the individual dose determined from this method will be 

overestimated, because at a minimum, intakes and doses from 12 radionuclides are considered. 

Annual organ doses were calculated as follows. For each radionuclide listed in Table 3.1.1-1, the 

annual organ doses from inhalation intakes were calculated with IMBA (Birchall 2002, 2003, 2003a, 

2003b). The absorption types for Table 3.1.1-1 radionuclides were selected based on the type that 

produced the largest non-metabolic annual organ doses. The ICRP 66 (ICRP 1994) respiratory tract 

model was used with the breathing rate of a standard worker. The assumed particle size distribution 

was 5 μm AMAD (default for worker exposures). The input summary is listed in Table 3.1.1-1. 

To facilitate entry of organ doses into the IREP computer code, Excel® workbooks (entitled 

"Maximum Internal Dose Calculation Workbook.xls” and “Maximum Internal Dose Calculation NonUranium Facility Workbook.xls”) was developed to create the IREP annual organ dose input data. 

Annual organ doses are summed over all radiation types and presented for each year that the 

covered employee was possibly exposed at the DOE site. Examples of annual organ dose output for 

the bladder are shown in Attachment A, Figures A.2.1-1 and A.2.1-2, for sites with and without 

reactors, respectively. In addition, Attachment A contains details of the dose calculations performed 

by the dose estimation Excel® spreadsheet. 

5.0 APPLICATIONS AND LIMITATIONS 

The following conditions must be met to apply this approach. 

•	 The covered employee’s initial hire date was after 1969. The covered employee’s initial hire 

date was after 1952 if their dosimetry records indicate that the covered employee did not 

participate in a bioassay program prior to 1970. 
 

Effective Date: 01/10/2004 Revision No. 01 PC Procedure No. ORAUT-OTIB-0002 Page 9 of 20 

•	 For Hanford employees, the covered employee’s hire date was after 1952. See Attachment B 

for discussion. 

•	 The covered employee was not included in a bioassay program or little or no detectable 

occupationally related activity was found in them or their samples. 

•	 The target organ must be listed in Table 3.1.1-4 or must be an organ whose dose is based on 

the highest non-modeled organ dose. 

•	 Employees who would have had no significant exposure to uranium can be assessed with the 

smaller uranium source term. 

•	 The employment must have been at a DOE site or national laboratory with an active radiation 

protection program. 

•	 Dose from Atomic Weapons Employers (AWEs) must be reconstructed by other methods. 

•	 Dose from Nevada Test Site may be assigned by this approach if the employee worked at 

NTS after 1971, was not involved in tunnel work and meets the other conditions herein. 

•	 Dose from tritium is to be considered separately. 
6.0 

Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 10 of 20 

REFERENCES 

10 CFR 835, Occupational Radiation Protection; Final Rule, December 14, 1993. 

42 CFR 82, 2002, Methods for Radiation Dose Reconstruction under the Energy Employees 

Occupational Illness Compensation Act of 2000, May. 

AEC (Atomic Energy Commission) Manual, Chapter 0524, Standards for Radiation Protection, AEC 

0524-01, November 8, 1968. 

Birchall, A, et al, 2003, IMBA ExpertTM-OCAS Edition software, Version 3.0.48, April. 

Birchall, A, James AC, IMBA-NIOSH (Phase 1) software, Version 1.0.42, 2002. 

Birchall, A, James AC, 2003a, IMBA ExpertTM-OCAS Edition software, Part 3 - Split SEEs, Version 

3.0.59. 

Birchall, A, James AC, 2003b, IMBA ExpertTM-OCAS Edition software, Part 3 - Split SEEs, Version 

3.0.63. 

EEOICPA (Energy Employees Occupational Illness Compensation Act), 2000, Public Law 106­

398,114 Stat. 1654, 1654A-1231, enacted as Title XXXVI of the Floyd D. Spence National 

Defense Authorization Act for Fiscal Year 2001, October. 

ICRP (International Commission on Radiological Protection), 1959, Report of Committee II on 

Permissible Dose for Internal Radiation, ICRP Publication 2, Pergamon Press, London, 

England. 

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, ICRP Publication 66, Annals of the ICRP Vol. 24 (1-3), Pergamon 

Press, Oxford, England. 

ICRP (International Commission on Radiological Protection), 1994a, Dose Coefficients for Intakes of 

Radionuclides by Workers, Replacement of ICRP Publication 61, ICRP 68, Annals of the ICRP 

Vol. 24 (4), Pergamon Press, Oxford, England. 

OCAS-IG-002, Rev.0, “Internal Dose Reconstruction Implementation Guideline,” August 2002. 

OCAS-TIB-008, Rev. 0, “Use of ICRP 66 to Calculate Respiratory Tract Doses”, Effective Date: 

9/29/03. 

Potter, C. A, 2002, “Intake Retention Fractions Developed from Models Used in the Determination of 

Dose Coefficients Developed for ICRP Publication 68 – Particulate Inhalation,” Health Physics, 

Vol. 83, No. 5, November. 

Recommendation of the National Committee on Radiation Protection, National Bureau of Standards 

(NBS) Handbook 69 (1959), Maximum Permissible Body Burdens and Maximum Permissible 

Concentration of Radionuclides in Air or Water for Occupational Exposure, Superintendent of 

Documents, U.S. Department of Commerce, U.S. Government Printing Office, Washington, 

D.C., June 5, 1959. 
Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 11 of 20 

ATTACHMENT A 

A.1 PURPOSE 

The purpose of this attachment is to provide a description of the Excel® workbooks entitled 

"Maximum Internal Dose Calculation Workbook.xls" and “Maximum Internal Dose Calculation NonUranium Facility Workbook.xls” that were developed to calculate annual organ doses in support of this 

Technical Information Bulletin (TIB). In addition, this attachment also provides documentation and a 

description of the IMBA (Birchall 2002, 2003, 2003a, 2003b) input and output files used to calculate 

and report organ doses and a description of the user workbook interface developed to assist the dose 

reconstructors in processing claims. 

The workbooks and the IMBA input and output files used in support of the Maximum Internal Dose 

Estimates for Certain DOE Complex Claims TIB are supplied as a folder, “ORAUT-OTIB-0002 Rev 00 

Attachment A,” on CD only. 

A.2 WORKBOOK DESCRIPTION 

The Excel® workbooks entitled "Maximum Internal Dose Calculation Workbook.xls" and “Maximum 

Internal Dose Calculation Non-Uranium Facility Workbook.xls” consist of multiple worksheets that are 

grouped according to function. These worksheets and their respective functions are described below. 

A.2.1 Dose Reconstructor Interface 

When a dose reconstructor uses the Maximum Internal Dose Calculation Workbook or Maximum 

Internal Dose Calculation Non-Uranium Facility Workbook, they will only see the worksheet entitled 

"UserPrompt."; all other worksheets contained within the workbook are hidden to assure the dose 

reconstructor enters the correct information in the appropriate location. This worksheet prompts the 

dose reconstructor for the following information: 

• First year of employment 

• Organ of interest 

• Reactor or non-reactor site 

Based on the responses of the dose reconstructor, when the "Open Template" button is pushed, the 

workbook will then calculate the annual organ doses and fill the data template that can be copied into 

an IREP template for the organ of interest. Figures A.2.1-1 and A.2.1-2 below show a typical output 

for a claimant whose first year of employment was 1975. The date of diagnosis of the cancer was 

1985, so the values after this year were manually deleted. The information provided in these 

templates is suitable for direct input into the IREP computer code (NIOSH-IREP 2003) used for 

estimation of probability of causation. Although only 10 years of organ dose information is shown in 

Figures A.2.1-1 and A.2.1-2, the actual template contains 65 years of organ doses to accommodate 

differing years of employment and cancer diagnosis dates. 

A.2.2 Maximum Permissible Body Burden 

The worksheet entitled MPBBs (Maximum Permissible Body Burdens) contains information related to 

the estimation of the derived intakes used for calculation of maximum organ doses. The assumptions 

used for selection of solubility types and intake factors used to derive assumed intakes are provided in 

the main body of the TIB. The worksheet also specifies the radionuclides appropriate for 
Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 12 of 20 

Figure A.2.1-1. Example output file of bladder dose for a covered employee who is assumed to have 

internal exposure beginning in 1975 at a non-uranium DOE complex site with a reactor and whose 

covered condition was diagnosed in 1985. 

Enter first exposure year in cell 

H3, for IREP input format 1975 

Copy internal dose data from the first year 

through the year of cancer diagnosis 

Exposure year Exposure rate Radiation type Dose distribution type Parameter 1 Parameter 2 Parameter 3 

1975 Chronic Alpha Constant 9.85E-02 0 0 

1975 Chronic Electrons E>15keV Constant 1.16E+00 0 0 

1975 Chronic Photons E>250keV Constant 7.21E-01 0 0 

1976 Chronic Alpha Constant 8.94E-02 0 0 

1976 Chronic Electrons E>15keV Constant 1.28E-01 0 0 

1976 Chronic Photons E>250keV Constant 1.20E-01 0 0 

1977 Chronic Alpha Constant 8.40E-02 0 0 

1977 Chronic Electrons E>15keV Constant 4.20E-02 0 0 

1977 Chronic Photons E>250keV Constant 2.93E-02 0 0 

1978 Chronic Alpha Constant 7.97E-02 0 0 

1978 Chronic Electrons E>15keV Constant 2.50E-02 0 0 

1978 Chronic Photons E>250keV Constant 9.48E-03 0 0 

1979 Chronic Alpha Constant 7.69E-02 0 0 

1979 Chronic Electrons E>15keV Constant 1.85E-02 0 0 

1979 Chronic Photons E>250keV Constant 4.36E-03 0 0 

1980 Chronic Alpha Constant 7.48E-02 0 0 

1980 Chronic Electrons E>15keV Constant 1.47E-02 0 0 

1980 Chronic Photons E>250keV Constant 2.55E-03 0 0 

1981 Chronic Alpha Constant 7.36E-02 0 0 

1981 Chronic Electrons E>15keV Constant 1.21E-02 0 0 

1981 Chronic Photons E>250keV Constant 1.66E-03 0 0 

1982 Chronic Alpha Constant 7.23E-02 0 0 

1982 Chronic Electrons E>15keV Constant 1.01E-02 0 0 

1982 Chronic Photons E>250keV Constant 1.13E-03 0 0 

1983 Chronic Alpha Constant 7.14E-02 0 0 

1983 Chronic Electrons E>15keV Constant 8.46E-03 0 0 

1983 Chronic Photons E>250keV Constant 7.94E-04 0 0 

1984 Chronic Alpha Constant 7.09E-02 0 0 

1984 Chronic Electrons E>15keV Constant 7.13E-03 0 0 

1984 Chronic Photons E>250keV Constant 5.67E-04 0 0 

1985 Chronic Alpha Constant 7.06E-02 0 0 

1985 Chronic Electrons E>15keV Constant 6.05E-03 0 0 

1985 Chronic Photons E>250keV Constant 4.12E-04 0 0 

consideration at both reactor and non-reactor sites. This worksheet is provided for information only 

and to document the derived intake calculations. The values contained within this worksheet are not 

linked to other worksheets in the workbook. 

A.2.3 Radionuclide Intakes 

The worksheet entitled "Intakes" lists the radionuclides (Column A, Rows 11 through 38), the IMBA 

absorption and radiation types (Columns B and C, Rows 11 through 38), and derived intake quantities 

for reactor and non-reactor sites (Columns D and E, Rows 11 through 38, respectively) used to 

calculate the maximum annual organ doses. The information related to the radiation-types, 65-year 

doses and radionuclide-specific dose fractions is provided for information only and to assist in the 

quality control process. Thus, this information-only data is not linked to other worksheets nor used 

directly by the worksheet for the organ dose calculations. 

The annual dose calculations are based on the intake values located in Column F, Rows 11 through 

38. These values are linked to the quantities in Columns D and E, Rows 11 through 38 for reactor 

and non-reactor sites, respectively by conditional statements based on user input from the 

UserPrompt worksheet, which is reproduced on the Intake worksheet at cell B2. The "Assumed 

Intake" values (derived intake quantities) in Column F, Rows 11 through 38 are then linked to the 
Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 13 of 20 

Figure A.2.1-2. Example output file of bladder dose for a covered employee who is assumed to have 

internal exposure beginning in 1975 at a non-uranium DOE complex site without a reactor and whose 

covered condition was diagnosed in 1985. 

Enter first exposure year in cell 

H3, for IREP input format 1975 

Copy internal dose data from the first year 

through the year of cancer diagnosis 

Exposure year Exposure rate Radiation type Dose distribution type Parameter 1 Parameter 2 Parameter 3 

1975 Chronic Alpha Constant 9.85E-02 0 0 

1975 Chronic Electrons E>15keV Constant 7.14E-01 0 0 

1975 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1976 Chronic Alpha Constant 8.94E-02 0 0 

1976 Chronic Electrons E>15keV Constant 6.84E-02 0 0 

1976 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1977 Chronic Alpha Constant 8.40E-02 0 0 

1977 Chronic Electrons E>15keV Constant 7.08E-03 0 0 

1977 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1978 Chronic Alpha Constant 7.97E-02 0 0 

1978 Chronic Electrons E>15keV Constant 1.00E-03 0 0 

1978 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1979 Chronic Alpha Constant 7.69E-02 0 0 

1979 Chronic Electrons E>15keV Constant 3.59E-04 0 0 

1979 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1980 Chronic Alpha Constant 7.48E-02 0 0 

1980 Chronic Electrons E>15keV Constant 2.51E-04 0 0 

1980 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1981 Chronic Alpha Constant 7.36E-02 0 0 

1981 Chronic Electrons E>15keV Constant 2.05E-04 0 0 

1981 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1982 Chronic Alpha Constant 7.23E-02 0 0 

1982 Chronic Electrons E>15keV Constant 1.72E-04 0 0 

1982 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1983 Chronic Alpha Constant 7.14E-02 0 0 

1983 Chronic Electrons E>15keV Constant 1.45E-04 0 0 

1983 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1984 Chronic Alpha Constant 7.09E-02 0 0 

1984 Chronic Electrons E>15keV Constant 1.23E-04 0 0 

1984 Chronic Photons E>250keV Constant 0.00E+00 0 0 

1985 Chronic Alpha Constant 7.06E-02 0 0 

1985 Chronic Electrons E>15keV Constant 1.05E-04 0 0 

1985 Chronic Photons E>250keV Constant 0.00E+00 0 0 

respective radionuclide sheets (at location E3) to provide the appropriate multiplier for the organ dose 

calculations (see discussion below). 

A.2.4 Radionuclides Important to Internal Dose 

Individual worksheets are included in the workbook for the 28 radionuclides determined to be 

important to internal dose (see Table 3.1.1-2 of the TIB). Each of these radionuclide-specific 

worksheets contains the derived intake values (cell E3), assumed intake adsorption type used for the 

IMBA runs (cell E2), radiation type (cell A1) and the IMBA type and version used for estimating organ 

doses (cell F1). 

On each of these radionuclide worksheets, results of the IMBA runs (i.e., annual organ dose in rem), 

for a one-nanocurie (nCi) acute inhalation intake of the respective radionuclide, are captured in cells 

B77 through AJ141 for each of the organs of interest (see below). Each of these values is multiplied 

by the derived intake quantity (cell E3) and the products are recorded in cells B6 through AJ70. 

These values represent the annual organ dose resulting from the derived intake values shown in 

Table 3.1.1-2 of the TIB and are summed, by radiation type in the "organ" worksheets that are used 

for the IREP inputs. The IMBA input and output for all radionuclides and organs of interest are 

reproduced on the Attachment A CD. 
Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 14 of 20 

A.2.5 Radiation Types 

Because the IREP computer code requires specification of radiation type, three worksheets were 

developed to add the IMBA doses from the radionuclide worksheets for each radiation type. The 

radiation type worksheets are entitled "Alpha,'" "Electron>15 keV," and "Photon>250keV." 

Radionuclides contributing dose to each of these radiation type worksheets are as follows: 

•	 Alpha – Cf-252, Am-241, Cm-244, Np-237, Pu-238, Pu-239, U-234, and U-238 

•	 Electron>15keV – Sr-90, Cs-137, I-131, Ce-141, Eu-154, Eu-155, Nb-95, Ru-103, Ru-106, Zr 

95, Y-91, Tc-99, and Pm-147 

•	 Photon>250keV – Co-58, Co-60, Mn-54, and Zn-65 

For each of the three radiation types, annual doses from the appropriate radionuclide worksheets 

(cells B6 through AJ70) were summed in cells B6 through AJ70 for each organ of interest. These 

annual doses summed by radiation type are then carried to the organ worksheets for preparation of 

the IREP input template. 

A.2.6 Organs of Interest 

Individual worksheets were included for each of the following organs of interest: 

•	 Adrenals 

•	 Esophagus 

•	 Pancreas 

•	 Thymus 

•	 Brain 

•	 G. Bladder 

•	 Skin 

•	 Thyroid 

•	 Breast 

•	 Heart 

•	 Spleen 

•	 U. Bladder 

•	 Colon 

•	 Muscle 

•	 Stomach 

•	 Uterus 

These organ worksheets arrange the organ doses by radiation type from all of the radionuclides of 

interest into a format suitable for direct entry into the IREP computer code (see Figures A.2.1-1 and 

A.2.1-2 above). The annual dose information by radiation type is linked from the radiation type 

worksheets and presented in columns B, C, and D; Rows 6 through 70, representing 65 years of 

exposure. These doses are then arranged in the IREP input form (shown in various colors) with three 

entries (one for each radiation type) for each year of exposure. These radiation- and year-specific 

annual doses (rem) are shown in Column J, Rows 6 through 200. The first year that dose is 

calculated is shown in cell H3 and is set based on input provided by the dose reconstructor through 

the UserPrompt worksheet. 
A.3 

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To setup the IREP input data, the dose reconstructor would copy cells F6 through L200 and paste 

these values into the identical cells in the IREP template. The actual number of yearly doses 

considered in the probability of causation calculation is set by the reconstructor in the "Number of 

Exposures" block in the IREP template. 

REFERENCES 

Birchall, A, et al, 2003, IMBA ExpertTM-OCAS Edition software, Version 3.0.48, April. 

Birchall, A, James AC, IMBA-NIOSH (Phase 1) software, Version 1.0.42, 2002. 

Birchall, A, James AC, 2003a, IMBA ExpertTM-OCAS Edition software, Part 3 - Split SEEs, Version 

3.0.59. 

Birchall, A, James AC, 2003b, IMBA ExpertTM-OCAS Edition software, Part 3 - Split SEEs, Version 

3.0.63. 

NIOSH-IREP, 2003, Interactive RadioEpidemiological Program version 5.2.1, National Institute of 

Occupational Safety and Health, Cincinnati, OH, May 
 

 

Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 16 of 20 

ATTACHMENT B 

Using Maximum Internal Dose Estimates from Technical Information Bulletin 0002 

for Hanford Claims from 1953 Forward 

B.1 PURPOSE 

The purpose of this attachment is to provide the basis for applying the assumptions, conditions, and 

results of ORAUT-OTIB-0002 to claims resulting from Hanford employment prior to 1970. 

B.2 BACKGROUND 

ORAUT-OTIB-0002 assigns intakes of a number of radionuclides based on the assumption that 

intakes resulting in greater than 10% of the Maximum Permissible Body Burden, as defined in ICRP 

Publication 2 (ICRP 1959) and NBS Handbook 69 (National Bureau of Standards 1959), would not 

have occurred without being detected by either workplace monitoring or bioassay or both. For sites 

with reactors, as would apply to Hanford, each claimant is assigned intakes of 28 radionuclides 

considered representative of potential sources of intake. In situations where more than one isotope 

might have been available as a potential source of intake the more restrictive isotope was assigned, 

for instance, 137Cs for mixtures of 134Cs and 137Cs, and 90Sr for mixtures of 89Sr and 90Sr. The intakes 

and resulting doses are considered conservative in that the risk of a worker truly incurring an 

undetected intake or series of intakes during a career involving all of the listed radionuclides at 10% of 

the MPBB for each is unlikely. 

The justification of extending ORAUT-OTIB-0002 to Hanford for hire dates earlier than 1970 requires 

an understanding the status of the radiation protection program for the earlier years. 

B.2.1 Brief Description of the Radiation Protection Program at Hanford 

Until 1965, there was one operating contractor at Hanford and therefore one radiation protection 

program. Herbert M. Parker, manager of the Health Instrumentation Division, and later manager of 

Hanford Laboratories, was the guiding force for radiation protection at Hanford in the 1940s and 

1950s. Dr. Parker was a leader in the field of health physics who pushed hard for continuously 

improving the science and the practice of health physics at Hanford and in the United States as a 

whole. Although managing radiation protection in general, Dr. Parker’s expertise was dosimetry, both 

external and internal. For example, Dr. Parker was a member of the NCRP subcommittee that 

developed the first set of Maximum Permissible Body Burdens (MPBBs) and Maximum Permissible 

Concentrations (MPCs). 

Jack Healy led developments in bioassay and internal dosimetry during this same time. Dr. Healy 

was also one of the foremost experts and leaders in these disciplines. For example, Dr. Healy was a 

member of the NCRP subcommittee that developed the second set of MPBBs and MPCs. 

Because of these men and others, radiation protection was a major concern at Hanford right from the 

start. Considerable research was conducted throughout most of Hanford’s history toward improving 

survey instruments, effluent cleanup mechanisms, monitoring instrumentation, workplace air sampling 

practices, and toward understanding the biokinetics associated with intakes of radionuclides. Some of 

these developments are described in Parker’s 1948 article in Advances in Biological and Medical 

Physics, “Health-Physics Instrumentation and Radiation Protection” (Parker 1948), and in his 1955 
Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 17 of 20 

article in Radiology, “Radiation Protection in the Atomic Energy Industry, a Ten-Year Review” (Parker 

1955). 

Nevertheless, the science and practice of monitoring for intakes of plutonium, uranium, and mixed 

fission and activation products were still in their infancy in the 1940s. Knowledge and skills 

associated with routine bioassay for plutonium and uranium were rapidly being developed and 

improved, and the scope of monitored workers was expanding. Limits were based on tolerance doses 

and tolerance air concentrations, which were forerunners of the MPBB and MPC concepts. But the 

tolerance doses were high by today’s standards: for instance, the external dose limit at Hanford was 

0.1 rep per day and later was revised to 0.3 rep per week [a rep was a radiation equivalent physical, 

defined as 83 erg/g energy deposition in tissue and later redefined as 93 ergs/g]. The internal dose 

limit for alpha-emitters was 0.01 rep per day (Cantril 1945). [Assuming a quality factor of 20, 0.01 rep 

approximately equates to 200 mrem per day.] Additionally the science of calculating intake or air 

concentration limits based on a limiting dose rate to a “significant organ” was based on very little 

biokinetic knowledge for radionuclides of concern associated with the Manhattan Project. 

Consequently, applying the default intakes from ORAUT- OTIB-0002 to the earliest years at Hanford 

is not recommended. 

B.2.2 Air Monitoring 

Air concentration limits at Hanford followed national guidance, and because of Drs. Parker and 

Healy’s involvement in setting national guidance, changes would have been implemented promptly, if 

not actually predating the publication of the guidance. For example, in a 1949 letter from C. M. 

Patterson of the Hanford Health Instrument Division to Lauriston S. Taylor of the National Bureau of 

Standards, in response to a request from Dr. Taylor, Dr. Patterson replied, [examples from the letter] 

“The permissible tolerances used at Hanford Works are as follows: . . . internal emitter limits – 0.3 

rep per week to the significant organ or whatever the National Committee on Radiation Protection 

recommends [italics not in original letter].” The italicized phrase was used in several places. And for 

drinking water, “mixed fission products – 0.1 μc/liter – to be changed when new figure provided by 

K.Z. Morgan’s Subcommittee.” (Patterson 1949) Dr. Morgan’s subcommittee was the NCRP 

Subcommittee on Permissible Internal Dose, of which Dr. Parker was a member, and which 

eventually produced the MPCs published in NBS Handbooks 52 and 69. 

Table B-1 compares the MPBBs and MPCs of key radionuclides provided as national guidance in 

1953 (NBS Handbook 52) and 1959 (NBS Handbook 69). This table is instructive because the 

MPBBs used in ORAUT-OTIB-0002 are directly linked to MPCs in NBS Handbook 69, and hence, to 

air concentration limits at Hanford from circa 1959. However, if one wishes to look back beyond 1959, 

the table shows what the MPCs were starting in 1953 and how those values compared to 1959 

values. 

Air monitoring at Hanford was based on gross alpha and gross beta measurements corrected for 

radon progeny. The beta MPC used at Hanford since on or before 1953 was based on the MPC for 

90Sr in the handbook applicable at the time. Because, as shown in Table B-1, the MPC for 90Sr 

actually increased between 1953 and 1959, using the MPBBs from the 1959 document is applicable 

and conservative for the 1953-1959 period. Similarly the alpha MPC at Hanford was 2 × 10-12 μCi/ml, 

based on 239Pu, which was unchanged from 1953. Areas with air concentrations exceeding either 

MPC were labeled as airborne contamination areas and required respiratory protection. So although 

the MPCs varied somewhat for the other radionuclides between 1953 and 1959, the 90Sr and 239Pu 

MPCs were always more limiting. Furthermore, these MPC values for 90Sr and 239Pu (and hence 
Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 18 of 20 

Table B-1. Comparison of MPBBs and Air MPCs between NBS Handbooks 52 

and 69. 

Radionuclide 

MPBB (nCi) MPC air (μCi/ml) 

Handbook 52 Handbook 69 Handbook 52 Handbook 69 

Pu-239 soluble 40 40 2 × 10-12 2 × 10-12 

Pu-239 insol. 8 NA 2 × 10-12 4 × 10-11 

U nat soluble 200 5 1.7 × 10-11 7x10-11 

U-238 soluble(a) 5 7 × 10-11 

U nat insol. 9 1.7 × 10-11 6 × 10-11 

U-238 insol. NA 1 × 10-10 

Am-241 56 50 3 × 10-11 6 × 10-12 

Sr-90 1,000 2000 2 × 10-10 3 × 10-10 

Cs-137 90,000 30,000 2 × 10-7 6 × 10-8 

Ce-144 5,000 5,000 7 × 10-9 1 × 10-8 

I-131 300 700 3 × 10-9 9 × 10-9 

Co-60 3,000 10,000 1 × 10-6 3 × 10-7 

Ru-106 4,000 3,000 3 × 10-8 8 × 10-8 

Y-91 15,000 5,000 4 × 10-8 4 × 10-8 

gross alpha and gross beta measurements on air samples) were continued via AEC Manual Chapter 

0524 and DOE Order 5480.11 into the 1980s. 

The point of this discussion is that the MPBB values used for ORAUT-OTIB-0002 were directly linked 

to the air monitoring program at Hanford since 1953, and that the air concentration limits and 

response to exposures potentially exceeding the limits were consistent from 1953 through to, at least, 

1989. 

B.2.3 Bioassay Monitoring 

Another key consideration in monitoring for intakes is the bioassay program. The plutonium bioassay 

program was well established and had years of operating experience at Hanford by 1953 (Bihl, 2003). 

In fact the plutonium urinalysis methodology changed little from March 1953 through September 1983, 

and the MDA was 0.05 dpm/24-hr sample. For type M plutonium, this MDA translates to a possible 

missed intake at 365 days post intake of 5.8 nCi total plutonium alpha compared to the ORAUT-OTIB0002 assumed intake of 160 nCi plutonium alpha. In addition the sensitivity of the program improved 

throughout the period due to increased use of fecal samples in response to potential intake incidents. 

Uranium bioassay (elemental analysis) was also well established by 1953 and was stable through 

about 1969 with an MDA of 4 μg/L. For type M recycled uranium, this MDA corresponds to a possible 

missed intake at 180 days post intake (sampling was conducted monthly for awhile and semiannually 

part of the time [Wilson 1987]) of 78 nCi total uranium alpha compared to the ORAUT-OTIB 0002 

assumed intakes of 55 nCi type F uranium for non-uranium facilities and 5500 nCi for uranium 

facilities. For type F recycled uranium, the corresponding possible missed acute intake at 180 days 

post intake is 163 nCi, but this is equivalent to 179 mg, which greatly exceeds the threshold for 

permanent renal damage (HPS 1995) and so is an unrealistic estimate. In fact, the 55 nCi intake in 

ORAUT-OTIB-0002 is overly conservative for type F uranium from the acute toxicity perspective, but 

most uranium intakes across the DOE complex were chronic. Uranium sampling was usually 

biweekly or monthly for workers exposed to highly soluble forms of uranium but some sampling was 

semiannual in the 1960s, and it is not know at this time under what conditions the semiannual 

sampling was allowed (for instance, it might have been just for workers exposed to insoluble forms of 

uranium). 
Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 19 of 20 

The largest MDA for the mixed fission product urinalysis method was 70 dpm/24-hr sample. When 

considered as pure type F 90Sr, this MDA corresponds to a possible missed intake at 365 days post 

intake of 1500 nCi compared to the ORAUT-OTIB-0002 assumed intake of 2,000 nCi. When 

considered as pure type M 144Ce, this MDA corresponds to a possible missed intake at 365 days post 

intake of 72,000 nCi compared to the ORAUT-OTIB-0002 assumed intake of 10,000 nCi. This implies 

that an intake of 144Ce at 10% of the MPBB would not have been detectable solely by an annual 

urinalysis. However an acute intake of 10,000 nCi of mixed fission products would have been 

associated with readily detectable contamination on the worker or in the air. For instance, breathing 

air at the beta concentration limit, 3 × 10-10 μCi/ml, for a week would have resulted in an intake of 14 

nCi. Allowing for a factor of 100 for possible non-representativeness of the air sample versus the air 

breathed by the worker still indicates that a 10,000-nCi intake would have been detected and resulted 

in prompt bioassay. 

The mixed fission product urinalysis did not detect intakes of isotopes of cesium or ruthenium or most 

activation products; however, these radionuclides are factored into the ORAUT-OTIB-0002 default 

intakes. 

B.2.4 Other Considerations 

A final consideration is how successful were the various monitoring programs at discovering and 

documenting intakes. In 1955 Parker reported in Radiology on the status of and improvements in 

radiation protection at Hanford for the preceding 10 years (Parker 1955). Included in this journal 

article was a table showing the number of internal depositions per year for 3 equal parts of the 

10-year period as follows: “positive cases per year, early period – 0; middle period --1; recent period 

– 30,” the recent period being about 1952-54. This statistic implies that either there were more 

workers being exposed in the latter period or detection capabilities had improved. The truth is 

probably both, but the point Parker was making was that the ability to detect and confirm intakes had 

improved considerably from 1944 to 1952. Since sometime prior to 1965 there was a policy that 

intakes leading to a body burden exceeding 5% of the MPBB were recorded in the worker’s radiation 

exposure file. Usually any indication of a possible intake was recorded along with the analysis leading 

to the 5% recording decision. Intakes well below the 5% criterion are listed in the records back to the 

1950s. These points - the improved capabilities of detection and the 5% MPBB policy - are not 

conclusive of themselves, but lend credence to the strength of the radiation protection programs by 

the early 1950s and the likelihood that intakes exceeding 10% of the MPBB would not have been 

missed. 

B.3 CONCLUSION 

•	 Air concentration limits of 3 × 10-10 μCi/ml total beta activity and 2 × 10-12 μCi.ml total alpha 

activity were based on the same MPBBs used by ORAUT-OTIB-0002 and were in use at least 

since 1953. These afforded adequate protection from acute intakes. 

•	 Routine bioassay from 1953 forward was also adequate for either acute or chronic intakes of 

plutonium or radiostrontium; was marginal for type M uranium, and would not detect an intake 

corresponding to 10% MPBB for type M 144Ce if the intake occurred 365 days prior to the 

bioassay. However, an intake of 10,000 nCi of 144Ce would have been associated with a large 

contamination incident and would have resulted in prompt bioassay. Same with large intakes 

of 141Ce, if someone had exposure to short-cooled fuel. 

•	 Records show that prior to about 1952 few intakes were being discovered or at least 

confirmed, but the radiation protection program showed a large increase in detections in about 
B.4 

Effective Date: 01/10/2004 Revision No. 01 Procedure No. ORAUT-OTIB-0002 Page 20 of 20 

1952. The records also show intakes well below 5% of the MPBB being logged starting at 

about that time. 

Hanford’s radiation protection program, consisting of air sampling using 90Sr and 239Pu MPCs for 

gross beta and gross alpha activities, respectively; workplace contamination monitoring and routine 

bioassay, was adequate from 1953 forward to ensure that intakes corresponding to 10% of the 1959 

MPBBs for 28 radionuclides described in ORAUT-OTIB-0002 would have been discovered and 

documented. The assigned dose from these 28 intakes is sufficient to overestimate doses for 

Hanford employees from 1953 forward, who meet all the other conditions described in ORAUT-OTIB0002. 

REFERENCES 

Bihl, D.E. 2003. “Technical Basis Document for the Hanford Site – Occupational Internal Dose.” 

ORAUT-TKBS-0006-5. Oak Ridge Associated Universities. 

Cantril, S.T. 1945. Letter to file, dated October 18, 1945, “Tolerance Limits.” 7-2602, also HW 4.159, 

Hanford Works. 

Health Physics Society (HPS). 1995. An American National Standard for Bioassay Programs for 

Uranium. HPS N13.22-1995. McLean, VA. 

International Commission on Radiological Protection (ICRP). 1959. Report of Committee II on 

Permissible Dose for Internal Radiation. ICRP Publication 2, Pergamon Press, London. 

National Bureau of Standards (NBS) 1953. “Maximum Permissible Amounts of Radioisotopes in the 

Human Body and Maximum Permissible Concentrations in Air and Water.” NBS Handbook 52 

(also referred to as NCRP Publication 11), NBS, Washington, D.C. 

National Bureau of Standards (NBS) 1959. “Maximum Permissible Body Burdens and Maximum 

Permissible Concentrations in Air and Water for Occupational Exposure.” NBS Handbook 69 

(also referred to as NCRP Publication 22), NBS, Washington, D.C. 

Parker, H.M. 1948. “Health Physics, Instrumentation, and Radiation Protection,” in Advances in 

Biological and Medical Physics, Volume I, Academic Press, New York. And reprinted in Health 

Physics, Vol. 38, No. 6, June 1960. 

Parker, H.M. 1955. “Radiation Protection in the Atomic Energy Industry, a Ten-Year Review,” 

Radiology, Vol. 65, No. 6, pp 903-11. Radiological Society of North America, Inc. 

Wilson, R.H. 1987. Historical Review of Personnel Dosimetry Development and its Use in Radiation 

Protection Programs at Hanford, 1944 to the 1980s. PNL-6125, Pacific Northwest Laboratory, 

Richland, Washington.